Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1

Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. The estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.

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Development of a Method to Assess Impact Hazards

Volume 3: Design and Analysis ◽

10.1115/pvp2015-45489 ◽

2015 ◽

Author(s):

Peter Gill ◽

Adam Toft

Keyword(s):

Structural Integrity ◽

Nuclear Reactors ◽

Primary Circuit ◽

Plant Safety ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Particle Hydrodynamics ◽

Smoothed Particle ◽

Circuit Components ◽

Safety Systems

The structural integrity of pressure retaining primary circuit components and of the containment boundary is of great significance for the safety justification of all nuclear reactors. In this regard, it is important to understand the vulnerability of safety related components to potential accidents. Whilst the direct consequences of pipe or pressure vessel failure, for example in terms of the extent of loss of coolant, may be tolerable, the indirect consequences of failure may not. A Loss of Coolant Accident (LOCA) may indirectly damage plant safety systems, including the containment boundary, due to the effect of missiles generated by the LOCA. This paper describes a study to develop a modern numerical modelling technique to estimate damage by missiles. Smoothed Particle Hydrodynamics (SPH) has been applied to simulate the acceleration of the missile due to the fluid jet.

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Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4037188 ◽

2017 ◽

Vol 3 (4) ◽

Cited By ~ 5

Author(s):

S. P. Saraswat ◽

P. Munshi ◽

A. Khanna ◽

C. Allison

Keyword(s):

Heat Flux ◽

Cooling System ◽

Heat Removal ◽

Hydraulic Modeling ◽

Carbon Fiber Composites ◽

High Heat Flux ◽

Vacuum Vessel ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Accident Scenarios

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

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Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

10.38032/jea.2020.02.004 ◽

2020 ◽

Vol 01 (02) ◽

pp. 53-60

Author(s):

Pronob Deb Nath ◽

Kazi Mostafijur Rahman ◽

Md. Abdullah Al Bari

Keyword(s):

Nuclear Reactor ◽

Cooling System ◽

Computational Study ◽

Reactor Core ◽

Primary Circuit ◽

Feed Water ◽

Pressurized Water ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

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AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model

Volume 6B: Thermal-Hydraulics and Safety Analyses ◽

10.1115/icone26-81886 ◽

2018 ◽

Cited By ~ 2

Author(s):

Samanta Estevez-Albuja ◽

Gonzalo Jimenez ◽

Kevin Fernández-Cosials ◽

César Queral ◽

Zuriñe Goñi

Keyword(s):

Cooling System ◽

Heat Removal ◽

Water Tank ◽

Local Pressure ◽

Water Storage Tank ◽

Passive Safety Systems ◽

Containment Model ◽

Loss Of Coolant Accident ◽

Ac Power ◽

Safety Systems

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

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